ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 36

DOE Review Release 01/27/2012

Document Title:

Exotic Radionuclides at the Savannah River 

Site

Document Number: ORAUT-RPRT-0054

Revision: 00

Effective Date: 01/20/2012

Type of Document: Report

Supersedes: None

Subject Expert(s): Edward D. Scalsky and Donald E. Bihl

Approval: Signature on File Approval Date: 01/19/2012

Edward D. Scalsky, Document Owner

Concurrence: Signature on File Concurrence Date: 01/19/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 01/19/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 01/19/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/20/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 2 of 36

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/20/2012 00 New document initiated in response to questions raised by SC&A 

that not all radionuclides have been identified and considered in the 

analysis of personnel exposures at the Savannah River Site. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Edward D. Scalsky.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 3 of 36

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 4

1.0 Executive Summary ................................................................................................................. 5

2.0 Introduction .............................................................................................................................. 5

3.0 Radionuclide Production at Savannah River ............................................................................ 5

4.0 Savannah River Laboratory Operations ................................................................................. 15

5.0 Bioassay Programs at SRS.................................................................................................... 16

6.0 Summary ............................................................................................................................... 22

References ........................................................................................................................................ 23

ATTACHMENT A, TECHNETIUM-99 ................................................................................................ 28

ATTACHMENT B, PROMETHIUM-147.............................................................................................. 30

ATTACHMENT C, BERKELIUM-249 AND FERMIUM-254 ................................................................ 32

ATTACHMENT D, THULIUM-171...................................................................................................... 34

ATTACHMENT E, AMERICIUM-242m .............................................................................................. 36

LIST OF TABLES

TABLE TITLE PAGE

3-1 Examples of general samples irradiated during the high flux demonstration............................. 7

3-2 Samples irradiated during the high flux program for selected months ...................................... 9

3-3 Radionuclides discussed for possible irradiation .................................................................... 12

3-4 Isotopes identified in SRS documents from 1954 through 1971 ............................................. 13

4-1 Positive bioassay results........................................................................................................ 16

5-1 Selected isotopes bioassay methods ..................................................................................... 19

A-1 Potential committed organ doses from 1.8 g 99Tc................................................................... 29

E-1 Activity of 242Cm per unit daily intake of 242mAm ...................................................................... 36
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 4 of 36

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL Argonne National Laboratory

BNL Brookhaven National Laboratory

d day

DNA delayed neutron analysis

DU depleted uranium

EU enriched uranium

g gram

GT Georgia Tech

HDEHP di,2ethylhexylophosphoric acid

HLC High Level Cave

hr hour

in. inch

keV kilo-electron volt

kg kilogram

LSC liquid scintillation counter

MCi megacurie

mg milligram

mm millimeter

MPBB maximum permissible body burden

MPLB maximum permissible lung burden

mrem millirem

nCi nanocurie

ORNL Oak Ridge National Laboratory

SRL Savannah River Laboratory

SRP Savannah River Plant

SRS Savannah River Site

TIOA tri-isooctlyamine

UCB University of California, LRL Berkeley

UCL University of California, LRL Livermore

UCLRL University of California Lawrence Radiation Laboratory

wk week

yr year

μg microgram
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 5 of 36

1.0 EXECUTIVE SUMMARY

The first reactor (R-Reactor) at Savannah River started operation on December 28, 1953 and shut 

down in June 1964. The K-, P-, and L-Reactors all started in 1954 and operated until 1988, while the 

L reactor shut down in 1968 and restarted in 1985. The C reactor operated from 1955 through 1985.

During the years the Savannah River Site (SRS) reactors were operating, there were many nuclides 

that were irradiated. Originally, the irradiations were in support of the nuclear weapons program. 

However, when military requirements slowed down, other missions were sought. Some missions 

included the production of transplutonium elements; development of isotopic power and heat power 

sources; and development of sources for food sterilization, medical uses, radiography, and irradiating 

samples for national laboratories, universities, and other countries in support of their research.

In light of the many radionuclides that resulted from these irradiations, questions have been raised 

that not all radionuclides have been identified and considered in the analysis of personnel exposures 

(SC&A 2010). According to the SC&A report, the main issues relate to the production, irradiation in 

reactors, and processing (if any) of target materials for the production of special radionuclides to 

clarify the dose reconstruction issues associated with them. SC&A classifies these special 

radionuclides as the “exotic” radionuclides that were produced in the C- and K-Reactors during the 

high flux years of operation (i.e., 1965 to 1970). This report attempts to identify the radionuclides, the 

bioassay methods that were used to evaluate the potential exposures to personnel handling and 

processing the radionuclides and, if possible, the approximate quantities that were available at SRS.

2.0 INTRODUCTION

In addition to the materials that were irradiated in the SRS reactors for the weapons program, a great 

deal of material was irradiated for use in commercial, industrial, and medical applications. This report 

discusses the nuclides that were irradiated and processed for various periods in the SRS facilities. 

The bioassay methods that were used to evaluate the potential exposure to these radionuclides are 

also discussed. 

3.0 RADIONUCLIDE PRODUCTION AT SAVANNAH RIVER

The term “exotic radionuclides” was used first in the SC&A report (SC&A 2010). SC&A states their 

memo concerns “exotic radionuclides” that are not explicitly addressed as matrix issues for which 

NIOSH is developing coworker models. Radionuclides excluded from analysis include 232Th, 241Am, 

244Cm, 252Cf, 237Np, fission and activation products,210Po, recycled uranium, natural uranium, 239/240Pu, 

and 241Pu. SC&A does include among the “exotic radionuclides” those materials that were irradiated 

for off-site customers during the high flux demonstration program. However, for purposes of this 

report any radionuclide with some exceptions (e.g., tritium, thorium) that were produced as a result of 

irradiation in the SRS reactors is considered to be an “exotic radionuclide”, some of which are only 

briefly mentioned.

The SC&A report (SC&A 2010) identified the following radionuclides of concern:

The alpha emitters were:

• Americium-243

• Protactinium-231

• Plutonium-238

• Plutonium-242

• Uranium-233
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 6 of 36

The beta emitters were:

• Cesium-137

• Europium-153

• Iodine-131

• Iridium-192

• Promethium-147

• Technetium-99

• Thulium-170

In the early 1950s, five heavy-water reactors (R, P, L, K, and C) were constructed at SRS. All five 

reactors were operating by 1955. Reactor shutdown began in 1964 with R-Reactor. C-Reactor was 

shut down in 1987; L-Reactor was shut down in 1968, restarted in 1985, and shut down again in 

1988. K- and P-Reactors operated with few interruptions until 1988. A restart of K-Reactor began in 

1991, but was not completed (Till et al. 2001). None of the reactors are currently operational.

Originally, the primary function of the SRS reactors was the production of plutonium and tritium for 

nuclear weapons. However, even in the early 1950s, the reactors were producing radioisotopes for 

nonweapons uses, which continued in the 1960s with the reduction of the Savannah River Site’s

military mission (Reed et al. 2002). Depleted uranium (DU), 237Np, and lithium targets were irradiated 

to produce 239Pu, 238Pu, and tritium, respectively. The targets were irradiated and transferred to the 

Tritium Facilities, H-Area Separations (238Pu), and F-Area Separations (239Pu) for processing. Spent 

reactor fuel was sent to H-Area Separations for processing. The reactors were also used to irradiate 

targets that produced transplutonic elements such as curium and californium.

Between 1955 and 1967 large quantities of special radionuclides were produced in the SRS reactors 

for various campaigns. For example, a series of campaigns produced 66 MCi of 60Co for Oak Ridge 

National Laboratory (ORNL), the U.S. Bureau of Mines, the U.S. Army Food Process Development 

Irradiator, Japan Atomic Energy Research Institution, Brookhaven National Laboratory (BNL), and 

Marine Products Development and 630 kg of 233U from irradiation of thorium were produced for the 

thorium breeder reactor program. In addition, 275 kg of 238Pu from irradiation of 237Np, 600 g of 210Po, 

1,700 Ci of 152Eu, and 370 MCi of 170Tm were produced. Other radionuclides produced during this 

time period and in subsequent years included more 210Po, 121Sn, 139Ce, 166Ho, and 41Ca for BNL, the

University of Illinois, and Denmark (Bebbington 1990).

From 1965 through 1970, C- and K- reactors were modified to be able to produce a very high flux of 

neutrons for production of unusual (e.g., 170Tm, 171Tm, 46Sc) radionuclide sources. The High Flux 

Demonstration Program began in C-Reactor in February 1965 and lasted 1 year. The High Flux 

Program in K-Reactor began in August 1969 to produce 252Cf. A peak flux of 7 × 1015 n/cm2

-sec was 

achieved and 2.1 g of 252Cf were produced before termination of the production campaign in 

November 1970. A second Californium Production Campaign was planned but never executed 

because the market for 252Cf never materialized as expected (Boswell 2000). The target material in 

the C- and K-Reactors consisted primarily of 242Pu. In addition, three 1-in.-diameter thimbles 

contained 243Am, 244Cm, and 150 target materials of 66 elements for nine universities and laboratories 

(Boswell 2000). Table 3-1 lists the target materials that were used to supply irradiated samples to 

Argonne National Laboratory (ANL), ORNL, University of California, LRL, Berkeley, and University of 

California, LRL, Livermore (Smith et al. 1965). There were additional targets; for example, one 

tritium-producing core (Mark VI) was designed with a removable inner lithium target so fuel could be 

irradiated to a very high exposure. The high exposure depletes 70% of the 235U and thus 

concentrates 236U, which is formed from every sixth neutron capture in 235U. Irradiation of the 236U 

forms 237Np, which is the target material for the production of plutonium (Boswell 2000). In addition, 

the irradiation of the target material 242Pu, after a succession of neutron captures and beta decays, 

successively produces other target materials – 242Pu, 243Pu, 243Am, 244Am, 244Cm, 245Cm, 246Cm, 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 7 of 36

247Cm, 248Cm, 249Cm, 249Bk, 250Bk, 250Cf, 251

Cf, and 252Cf (Reed et al. 2002). There were also special 

irradiations during both the high-flux charge and the Curium I campaign to produce test quantities of 

170Tm, 171Tm, 46Sc, and 204Tl.

Table 3-1. Examples of general samples irradiated during the high flux 

demonstration.

a,b

Sample Lab Sample Lab Sample Lab

Si-30 A Yb-168 OR Os UCL

S-33 OR Tm-168 UCB, UCL W UCL

Ca-40 OR, A Er-170 UCB Re UCL

Ca-48 A Yb-176 A, UCB, UCL Be A

Fe-54 OR Lu-176 UCB Be3N2 OR

Fe-58 OR, A, UCB Hf-177 UCB Xe UCL

Ni-58 OR, A Hf-179 A, UCL, UCB TiO2 A

Ni-62 OR Hf-180 A, UCL, UCB

Nb-93 OR Ta-180 UCB

Nb-96 A Ta-181 UCB

Ag-107 UCB W-186 A, UCL, UCB Additional isotopesc

Ag-108 UCB Re-187 UCL, UCB Zn-67 A

Sn-118 OR, UCB, A Os-180 UCB Te-124 A

Sn-124 A Os-182 A, UCL, UCB Lu-175 UCL

Ba-130 UCB Bi-209 UCB Be-9 A

Sm-144 OR, UCB Am-241 A, UCL

Nd-146 OR Cm-244 A, UCL, UCB Sn-120 A

Nd-150 A Bk-249 A

Sm-150 UCB, UCL Cf-252 A, UCL, UCB Cd-160 UCL

Sm-154 OR Es-254 UCB Am-243 UCL

Dy-156 UCB, UCL LiCl UCB

Dy-158 A MgO+Cr-50 UCL

Tb-158 UCB, UCL KPO3 UCL

Gd-160 A, UCL, UCB Ar UCL

Er-162 UCB, UCL RbCl OR

Dy-164 A, UCB Hf A

Ho-165 UCB, UCL HFO2 UCL

a. A=Argonne National Laboratory; OR=Oak Ridge National Laboratory; UCB=University of 

California, LRL, Berkeley; UCL=University of California, LRL, Livermore.

b. Source: Smith et al. 1965.

c. Other isotopes from Table 3-2.

Some development work was done mostly with cold Tm2O3 (DuPont 1974, p. 12) but a little work was 

done with 170Tm, 171Tm, 46Sc, and 204Tl (DuPont 1974, pp. 7–8). Scandium-46 is a gamma-emitter. 

Thallium-204 is a beta-only emitter. Thulium-170 and -171 have low-yield, low-energy gammas that 

might show up in chest counting if someone was looking closely, but mostly are beta emitters. 

Thulium oxide was characterized as target material for irradiation to form 170Tm, with potential 

applications for short-term missions in space (128-day half-life). As with cobalt, the irradiated product 

is useful without chemical processing. Conditions were established for fabricating wafers, and the 

properties of Tm2O3 were studied to prepare and irradiate it to yield a radioactive product suitable for 

use in heat sources. There was some activity that involved breaching of the source or doing 

chemistry on the 170Tm and 171Tm. These were small research and development experiments before 

actual large-scale production of the sources. There was some breakage of 170Tm oxide wafers in the 

early experiments. Metal targets did not have that problem. There was no evidence of ongoing,

large-scale breaching or chemistry on the 170Tm or 171 Tm sources. Thulium-170 was mentioned as a 

possible heat source early in the program; it emits a gamma ray with about 3% yield. Thulium-171 

only emits a beta in sufficient abundance for bioassay.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 8 of 36

SC&A mentioned 147Pm and 192Ir as two radionuclides of concern (SC&A 2010). The National 

Aeronautics and Space Administration’s interest in aged 147Pm as a source of Isotopic heat for space 

missions led to a study of the quantity and quality of 147Pm that could be produced annually at the 

SRP. The recovery of 147Pm from current high-activity waste was also studied. The major source of 

147Pm would be fission product waste; the quantity available after aging to allow for decay of 

associated 148Pm would be about 1,700 W(t) per reactor-year of operation. Product quality from 

current waste, as represented by a low ratio of 148Pm to 147Pm, would presumably be the same as that 

produced at the Hanford Site (about 5 × 10-7). Promethium-148 emits gamma rays that could require 

extra shielding. Irradiation of naturally occurring neodymium would produce only 50% as much 147Pm 

(DuPont 1974). No further mention of 147Pm has been found in this historical reference. DuPont

(1974) mentions that iridium was experimented with but indicates that this experimentation was with 

several inactive iridium capsules. The report makes no mention of 192Ir.

Table 3-1 shows a list of materials that were irradiated for the AEC laboratories only, i.e., Argonne 

National Laboratory (ANL), Oak Ridge National Laboratory (ORNL), University of California, Berkeley 

(UCB), and the University of California, Livermore (UCL). Some of the materials were irradiated for 

the duration of the high-flux campaign (i.e., from February to November 1965). Others were irradiated 

for single fuel cycles of about 1 week (Smith et al. 1965). Table 3-1 does not include isotopes that 

were irradiated for BNL, the University of Illinois, or Denmark, such as 121Sn, 138Ce, and 139Ce, or 

those irradiated for all off-site customers.

Table 3-2 lists some of the specific target elements and compounds that were irradiated for outside 

customers during the months and years indicated. Many of the samples are repeated in subsequent 

months; for example, March and May 1966 and March 1967 have many repeat samples. There were 

five shipments during January 1966. The 257Fm to LLNL and the 2 μg of 252Cf to ANL are listed in the 

table. The other shipments, not listed, included 380 mg of iridium to ORNL, 200 mg of calcium to 

Berkeley, and 1 μg of 253Es to ANL.

The procedure for handling these samples is described by Gray (2006) as follows:

All samples containing the material to be irradiated for off-site customers were in 

welded closed aluminum cans. They came to SRP in that condition. They were placed 

in a basket and the basket lowered into the spent fuel storage basin where the deep 

water provided shielding for handling radioactive components removed from the 

reactor. The research samples were placed in small perforated tubes about 1 inch in 

diameter; maybe five or so samples per a 12" tube. The tubes were then placed in a 

larger (about 1 3/8ths inch diameter) perforated tube about 21 feet long, called a 

thimble. The reactor had many of these thimbles for general use with safety rods, 

instruments, etc. Three of them near the center of the reactor were dedicated to my 

[Gray] use for research samples. When it came time for either insertion or removal (or 

both) of a sample, the thimble was removed by remote machine from the reactor and 

placed in the basin. It was moved remotely under water to a location where I could 

have it turned upside down and the tubes unloaded onto a well-lit work table about 12 

feet down. I sorted through the tubes and found the one from which a sample would 

be unloaded (or a new one loaded); all under water. The process was reversed to put 

the whole thing back together and return the loaded thimble to the reactor for the next 

cycle. Of the 15-18 hours for shut down, I had about 3-5 hours in the middle of it to do 

my work.

A removed research sample, without any opening of the welded-closed aluminum 

cans, was placed directly in a shielded shipping cask. The cask lid was put in place 

and sealed (or locked) down for transport. As the cask was removed from the basin it
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 

9 of 36

Table 3-2. Samples irradiated during the high flux program for selected months.

a

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

ORNL 280 mg natural 

iridium

SRP-SRL 1.4 g Co-59 UCLRL 11 mg Ag-107 UCLRL 11 mg Ag-107 UCLRL 11 mg Ag-107

SRP-SRL 1.4 g Co-59 ANL 40 mg Cm-244 UCLRL 1.52 g LiCl UCLRL 1.52 g LiCl UCLRL 1.52 g LiCl

ANL 40 mg Cm-244 ANL 2 μg Cf-252 UCLRL 217 mg MgO 

and Cr-50

UCLRL 217 mg MgO 

and Cr-50

UCLRL 217 mg MgO 

and Cr-50

ANL 2 μg Cf-252 UCLRL 240 mg 

Cm-244, 

20 mg Cm-246

UCLRL 10 mg Fe-58 UCLRL 10 mg Fe-58 UCLRL 10 mg Fe-58

UCLRL 400 mg Cm-244 ANL 77.2 mg Be-9, 

0.6 mg Ca-48, 2 

mg S-30

UCLRL 2 mg Dy-156 UCLRL 2 mg Dy-156 UCLRL 2 mg Dy-156

UCLRL 450 mg 

Cm-244; 

40 mg Cm-246

ANL 3.95 mg Fe-58, 

1.4 mg Sn-124, 

5.26 mg Hf-180

UCLRL 9 mg Hf-179 UCLRL 9 mg Hf-179 UCLRL 9 mg Hf-179

UCLRL 60 mg Cm-244; 

25 mg Cm-246

UCLRL 1.9 mg Cf-252 UCLRL 10.3 mg Hf-180 UCLRL 10.3 mg Hf-180 UCLRL 10.3 mg Hf-180

ANL 240 mg 

Cm-244; 

20 mg Cm-246

UCLRL 231 mg Gd-160 UCLRL 58 mg Sm-150 UCLRL 58 mg Sm-150 UCLRL 58 mg Sm-150

ANL 77.2 mg Be-9; 

0.6 mg Ca-48; 

2 mg Si-30

UCLRL 48 g Bi-209 GT 200 mg 27% 

Pb-204 metal, 

1 mg 99.7% 204PbO

GT 200 mg 27% 

Pb-204 metal, 

1 mg 99.7% 204PbO

GT 200 mg 27% 

Pb-204 metal, 

1 mg 99.7% 204PbO

UCLRL 3.95 mg Fe-58; 

1.4 mg Sn124; 

5.26 mg Hf-180

UCLRL 48 gm Bi-209 BNL 0.5 g C 

(diamond in 

graphite)

UCLRL 1.5 mg Pu-240, 

0.1 mg Pu-244

UCLRL 1.5 mg Pu-240, 

0.1 mg Pu-244

ANL 1.9 mg Cf-252 UCLRL 99 mg Yb-176 ANL 40 mg Cm-244 UCLRL 48 g Bi-209

(2.5 in.)

UCLRL 48 g Bi-209

UCLRL 0.8 g Hf-180 UI 8.3 mg 

Ba(NO3)2, 

12.5 mg CeO2, 

600 mg Sn-118

UCLRL 1.5 mg Pu-240, 

0.1 mg Pu-244

UCLRL 48 g Bi-209

(2.54 in.)

UCLRL 48 g Bi-209

UCLRL 231 mg Cd-160 UCLRL 1 g Am-243c UCLRL 48 g Bi-209 UCLRL 10.8 mg Hf-177 UCLRL 10.8 mg Hf-177

UCLRL 48 g Bi-209 UCLRL 1 g Am-243 UCLRL

ANL 

240 mg 

Cm-244; 

20 mg Cm-245

UCLRL 10 mg Os-192 UCLRL 10 mg Os-192

UCLRL 48 g Bi-209 UCLRL 5.7 g Lu-175 UCLRL 48 g Bi-209 UCLRL 10.9 mg Er-162 UCLRL 10.9 mg Er-162
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 10 of 36

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

UCLRL 99 mg Yb-176 ORNL 216 mg Sm2O3 

(enriched to 

94.5% Sm-144) 

and 7.828 mg 

type 302 SS 

monitor

UCLRL 10.8 mg Hf-177 UCLRL 9.6 mg W-186 UCLRL 9.6 mg W-186

SRP 13.47 mg Al ORNL 700 mg gold foil 

and 7.61 mg 

type 302 SS 

monitor

UCLRL 10 mg Os-192 UCLRL 9.74 mg Gd-160 UCLRL 9.74 mg Gd-160

UI 8.3 mg 

Ba(NO3)2; 

12.5 mg CeO2; 

600 mg Sn-118

ORNL 418 mg La2O3

and 7.718 mg 

type 302 SS 

monitor

UCLRL 10.9 mg Er-162 UCLRL 9.74 mg Re-187 UCLRL 9.74 mg Re-187

ORNL 280 mg natural 

iridium

ANL 1 gm Sn-120 UCLRL 9.6 mg W-186 UCLRL 7.65 mg Lu-176 UCLRL 24 mg Sm-144

UCLRL 1 g Am-243 UCLRL 40 mg Ba-130 

as BaCO3

UCLRL 9.74 mg Gd-160 UCLRL 24 mg Sm-144 UCLRL 50 mg Sn-118

UCLRL 1 g Am-243 UCLRL 40 mg Ba-130 

as BaCO3

UCLRL 9.74 mg Re-187 UCLRL 9.6 mg Ag-109 GT 77 g lead

UCLRL 1 g Am-243 UCLRL 40 mg Ba-130 

as BaCO3

UCLRL 24 mg Sm-144 UCLRL 50 mg Sn-118 SRL 1 g Er2O3

enriched in 

Er-170

UCLRL 1 mg Hg-203 

(as nitrate)

UCLRL 1.5 mg Pu-240, 

0.1 mg Pu-244

UCLRL 9.6 mg Ag-109 ORNL 100 mg Te-128 

(99.16% 

enriched)

ANL 60 mg Te-126 

enriched 

(98.69%)

SRL 90 mg 146Nd2O3

(enriched)

BNL 0.5 g C 

(diamond in 

graphite)

UCLRL 50 mg Sn-118 GT 77 g lead Livermore 460 mg Cm244, 

90 mg Cm-246, 

9 mg Cm-248

BNL 0.5 g C 

(diamond in 

graphite)

ORNL 100 mg Te-128 

(99.16% 

enriched)

NC 500 mg SnO2 

enriched 99% 

Sn-120

BNL 0.5 g C 

(diamond in 

graphite)

ORNL 100 mg 128Te 

(99.16% 

enriched)

Livermore 300 μg Cf-252

ANL 40Ca, 42Ca, 58Ni, 

62Ni,

GT 77 g lead ANL Er-162, Yb-176, 

Mo-92
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 11 of 36

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

ANL Zn-67, Te-124, 

Pb-204d, g

ANL Fe-54, Yb-168, 

Cr-50, Os-192

ANL Se, Yb-176 g ANL Nd-150, Ti-50, 

W-186, Yb-176

ANL Nd-150,

Os-190g

ANL Nb, Ce-136, 

Cf-249, N-236p

ANL U-236, Pa, Ni g ANL 90 mg Cm-244, 

30 mg Cm-246

ANL Np-236, Cf-249, 

Th-229g

UCLRL 5 mg Ln-176 

(70% enriched)

UCLRL 200 mg calcium ORNL 103.2 mg 

186WO3 (97.2% 

enriched)

UCLRL 0.7 mg platinum 

metal

UCLRL 1 mg Hg-202 in 

quartz capsule

UCLRL Fm-257 rabbit 

sample in Al 

capsule (10 μg)

ORNL 8.5 mg type 302 

SS monitor

ORNL 150 μg 236U2O3, 

3 - 8.5 mg type 

302 SS monitor

a. ANL-Argonne National Laboratory, ORNL-Oak Ridge National Laboratory, UCLRL-University of California Lawrence Radiation Laboratory, BNL-Brookhaven National 

Laboratory, GT-Georgia Tech, UI-University of Illinois, SRL-Savannah River Laboratory, SRP-Savannah River Plant. 

b. DuPont (1965a). 

c. DuPont (1966a). 

d. DuPont (1966c). 

e. DuPont (1965b). 

f. DuPont (1967a). 

g. The sample mass and chemical form of one or more nuclides were unreadable in the referenced document. 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 12 of 36

was washed down. Then it was placed on a truck for transport, usually to the Atlanta 

airport, for a flight to the outfit that sent it to us originally for irradiation. Many samples, 

because of short half lives of the desired product, were returned in well less than 24 

hours. No samples were opened at SRP [Savannah River Plant]. 

As discussed by Gray (2006), all samples that were to be irradiated for external customers were in 

welded closed aluminum cans. They came to SRS in that condition. After irradiation they were 

prepared under water for shipment, so the potential for intake was minimal. Similarly, because the 

sources were inspected and loaded into casks under water, the potential for external exposure was 

reduced. As noted above, Table 3-2 lists specific samples that were irradiated for the sponsors listed 

in the odd numbered columns for the month and year indicated. As can be seen in the table, the 

samples are generally small for research purposes at the sponsor’s facilities. However, there are five 

samples in the 5-month period listed in the table that were transferred to SRS-SRL. These samples 

would have been handled using SRS procedures. Based on the limited number of radionuclides 

transferred, it would appear that such irradiations were not routine. 

In addition to the radionuclides and compounds listed in Tables 3-1 and 3-2, a request was made of 

the Isotopes Development Center at SRS to look at the possibility of using Site irradiation facilities as 

fully as possible for the production of unusual or quite large amounts of isotopes for the potential 

benefit to research and commerce. In response to that request, Rupp (1969) discussed some of the 

characteristics of the radionuclides in Table 3-3, the target materials or compounds that would be 

used to produce some of them, and difficulties that might be encountered. He noted that some of 

these radionuclides were already being produced. As noted above in this section, these radionuclides 

have been identified by SC&A (2010) as being of concern for dose reconstruction. The entire list 

discussed in Rupp (1969) is included in Table 3-3. No documentation has been found indicating that 

program was ever pursued.

Table 3-3. Radionuclides discussed for possible irradiation (Rupp 1969). 

Plutonium-238 Thulium-170 and 171 Calcium-41 Barium-133 Samarium-154

Americium-242m Potassium-40 Iron-55 Lanthanum-137 Samarium-155

Cesium-134 Argon-38 (stable) Nickel-59 Terbium-157 Europium 155

Europium-152 Chlorine-36 Cadmium-109 Holmium-166m

Europium-154 Argon-39 Cadmium-113m Samarium-151

In addition to the information above, a review of monthly reports was conducted to identify the 

radionuclides that were being irradiated or processed at the SRS from 1954 through 1971. Table 3-4 

lists the results of this review (Isotope Charts 2011). Some of these isotopes were part of the 

production process and many were shipped to off-site customers.

The cells containing an x indicate that the isotope was present several times during the year. Some 

isotopes were present in every month of the year, others only during several months. Those identified 

by an L or M indicate they were present on a limited basis during 1 or 2 months during the year. The 

letters represent the potential for exposure to that radionuclide [e.g., either low (L) or medium (M) 

during that period]. This represents the highest estimated exposure potential for those radionuclides. 

Because the radionuclides indicated by an x appeared many times during the year, an indication of 

their exposure potential could not be identified in the table. Most were in the low to medium range but 

others, such as 241Am, 252Cf, 244Cm, and 237Np, were often listed as a high exposure potential. The # 

symbol indicates the isotopes were identified either through existing SRS calculations or in plans. 

There are 4 radionuclides added to the table that were not identified in the monthly reports but were 

identified in History of the Savannah River Laboratory, Volume IV – General (DuPont 1984a) and 

were observed in the bioassay results. These are 106Ru, 233Pa, 95Zr, and Zr/Nb-95 and are included 

for completeness.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 13 of 36

Table 3-4. Isotopes identified in SRS documents from 1954 through 1971. 

Isotope 1954 1955 1956 1957 1858 1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970 1971

Am-241 x x x x x x

Am-242ma #b Mc

Am-243 x x x x x x x x x x

Ala Ld

Ba-133a L

Bi-209a L

Bk-249 x x x x x x

Bk-250a L M

Ce-144a x x x x x

Cf-252 x x x x x x x x x

Cm-244 x x x x x x x x x x x

Cm-248f f

Co-59 x x

C0-60 x x x x x x x

Cs-137a x x x

Er-170 x x x

Es-253 x x x x

Es-254 x x x x

Es-255 L

Eu-152a L L

Eu-154a L L

Fm-254a M

Flux wire L

Ir-192a L

Nd-148a L

Np-237a x x x x x x x x x x x x x x x

Np-238a M

Np-239 L

Pa-231a M

Pa-233e e

Pm-147a x x x x x x x

Po-210a x x x x x x x

Pu-238a x x x x x x x x

Pu-239a x

Pu-240a x x x x x x # #

Pu-241a x

Pu-242a x x x x x x x x

Ru-106a,e e
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 14 of 36

Isotope 1954 1955 1956 1957 1858 1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970 1971

Sb-124a L

Sb-125a L

Sc-46a L L

Si (quartz)a L

Taa L

Tc-99a L

Tl-204a L L

Tm-170a x x x x x

Tm-171a x x x x x

U-233a x x x x x x x x

Y-91a L

Zr-95e e

Zr-95-Nb-95e e

Thoriuma x x x x x x x x x x x x x

a. These isotopes are not included in Table 3-2.

b. # Identified in calculations or plans.

c. M Represents an estimated moderate potential of exposure.

d. L Represents an estimated low potential of exposure. 

e. Not identified in the review of the monthly reports but was identified by the SRL bioassay program (DuPont 1984a. 

f. During separation and purification of transcurium nuclides, 1.7 μg 248Cm was obtained from a filter from which 14 μg 252Cf had previously been obtained. 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 15 of 36

4.0 SAVANNAH RIVER LABORATORY OPERATIONS

The Savannah River Laboratory (SRL) (Building 773-A) and related organizations began organizing in 

July 1953. Over the next couple of years the Engineering' Services that were provided in support of 

research and development personnel included operation of all building auxiliaries and waste disposal 

facilities, a glass shop, instrument pool, and a supply room. Close liaison was established with Works 

Engineering Department personnel to coordinate all maintenance, machine shop, and project work 

required for the Savannah River Laboratory. As radioisotopes were brought into 773-A to start 

operations in radiochemical laboratories, the Radiation Control group assisted in establishing 

necessary procedures and handling techniques.

Operations formally accepted occupancy of the cave wing, 773-A, Part II, on December 7, 1953 and 

all construction work was completed in February 1954. Experiments with radioactive materials began 

in the caves in July 1954. In October 1954 the first plutonium produced at the Savannah River Plant 

was isolated in the caves.

A review of the operations in the SRL was conducted because that was where the products of reactor 

operations were evaluated. While the SRL did not evaluate the radionuclides that were shipped to offsite customers, they did handle, evaluate, experiment with, and store “production” radionuclides such 

as 244Cm, 252Cf, 60Co, etc. Reactor incidents are not covered in this report because they would 

primarily involve reactor functionality, such as fuel failures, valve failures, etc., rather than exposure to 

the handlers of the transfers of highly radioactive radioisotopes, which was, of necessity, performed 

under water due to the high exposure potential. There were a considerable number of potentially 

serious incidents in the SRL, as discussed below. Several of the incidents have been investigated 

and have an associated Special Hazard Investigation report (Dosimetry Special Hazard Incident 

Investigations #’s 200-299), while the nature of many of the incidents has not been identified. 

However, there was an extensive radiation control program in terms of identifying potential problems 

and steps taken to evaluate any internal or external exposure as a result of these incidents.

Radiation Control was established in Technical Division facilities so all the survey, monitoring, and 

"containment' services usually assigned to Health Physics could be provided in the Technical Division 

organization, together with a wholly integrated approach and direct responsibility (DuPont1984a). The 

containment service function required close liaison work with technical personnel in the design and 

construction of the special equipment required to contain the radioactive materials completely during 

experimental work. This function was performed by the Apparatus Development Group, a subgroup 

of Radiation Control (DuPont 1984a). As radioisotopes were brought into Building 773-A to start 

operations in radiochemical laboratories, the Radiation Control Group assisted in establishing 

necessary procedures and handling techniques. Major effort was directed to extensive training of 

laboratory personnel and personnel of assigned groups. The fundamentals of radiation and 

contamination control and aspects of personal protection were stressed. By December 1955, 

procedures basic to the program of personnel and area radiation protection were essentially complete 

(DuPont 1984a).

Much of the work in the SRL was carried out in the High-Level Caves, which were designed to 

perform chemical and metallurgical operations on highly radioactive materials. As indicated above, 

operations formally accepted occupancy of the cave wing, 773-A, Part II, on December 7, 1953. In 

October 1954 the first plutonium produced at the Savannah River Plant was isolated in the caves to 

provide basic data to substantiate and firm up calculations of reactor performance. Operations in the 

caves continued at the maximum obtainable rate from that date.

The following information presents a synopsis of the internal exposure history of the SRL from 1953 

through December 1972, which encompasses the high-flux operational period. The information for 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 16 of 36

these years is taken from the History of the Savannah River Laboratory, Volume IV – General

(DuPont 1984a). 

In the period from July 1953 to the early part of 1960 evaluation of internal doses was based on invitro analysis. In 1960 the SRS whole body counter was completed and provided another means of 

assessing internal doses. From 1953 through 1972 there were thousands of urine samples and whole 

body counts but relatively few documented uptakes in the SRL that resulted in significant doses. 

There were no significant biological uptakes from 1953 through 1959 and none identified in 1963 and 

1968. Table 4-1 shows the number of positive bioassay results that have been documented in 

DuPont (1984a). Not all positive bioassay results were confirmed to be significant uptakes. 

Two Special Hazards Investigation (SHI) reports have been identified during this period. Special 

Hazards Investigation No. 273 reported a general service operator received a positive uptake of 244Cm 

when he released airborne activity to the decontamination room environment while removing solid 

radioactive waste from the Decontamination Chamber in Building 773-A, Decontamination Facility, 

Room C-003 (DuPont 1970a). Special Hazard Investigation No. 283 reported an Analytical Chemistry 

Division Chemist assimilated some curium while removing samples from storage in Building 773-A, 

Room B-15 (DuPont 1971a). 

Table 4-1. Positive bioassay results. 

Year

Number of 

positive results Radionuclide % MPBB

1960 5 Not identified < 1

1961 3 Not identified < 1

1962 51 Not identified < 1

1964 1 Pa-233 < 1.5

1 Ce-144 < 1.5

Pu < 1.5

1 Not identified Not identified

1965 1 Am-Cm 30a

1 Not identified < 5

1966 1 Zr-95 3.2

Ru-106 1.2

4 Pa-231 < 1

1967 1 Ru-106 4.2

Ce-144 1.4

Zr-Nb-95 0.36

1969 11 Cm-244 Not identified

1970 31 Cm-244 Not identified

1 Cm-244 SHI No. 273b

1971 21 Not identified Not identified

1 of the 21 Cm-244 SHI No. 283c

1972 4 Not identified Not identified

2 Cf-252 Not identified

a. No Hazard Investigation Report was found for this uptake. 

b. DuPont (1970a). 

c. DuPont (1971a). 

5.0 BIOASSAY PROGRAMS AT SRS

Workers at SRS were monitored for internal deposition of radionuclides from the beginning of 

operations. The methods used for evaluation of potential intakes included analysis of the amount of 

radioactivity excreted in the urine or feces (in vitro bioassay) and the measurement of the photon 

radiation emitted from the body by chest counting or whole-body counting (in vivo bioassay).
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 17 of 36

In the beginning of operations at SRS, the only method available to evaluate the potential internal 

exposure of personnel was in vitro bioassay, primarily analysis of urine samples. This method was 

typically a gross beta activity analysis although a gross gamma analysis was included from at least 

1959 (Boni undated). Over the years, the bioassay analyses methods changed as new techniques 

were developed. 

In Vitro Bioassay

In vitro bioassay methods changed considerably over time as new techniques and analyses were 

developed and refined. Some of the methods used for analysis of some of the radionuclides are 

briefly discussed below.

Before completion of a whole-body counting facility in 1960, only in vitro bioassay, primarily urinalysis, 

was performed. In vitro urine bioassay for nuclear byproducts performed during this period was 

typically a gross beta activity analysis and/or gross gamma analysis. Since that time, in vitro bioassay 

analysis has become more radionuclide-specific, and can be divided into the following categories:

• Plutonium urinalysis 

• Uranium urinalysis

• Trivalent actinide urinalysis

• Neptunium urinalysis

• Fission product urinalysis

• Fecal bioassay

Plutonium Urinalysis

The first procedure for plutonium urinalysis at SRS was implemented in 1954 and used until 1959. 

Changes in the processes to prepare the solution for counting were made but the method for counting 

the alphas tracks on the nuclear track emulsion (NTA) remained the same until 1966. In 1964 solidstate surface barrier detectors (gross alpha counting) replaced the NTA detectors. In the early 1980s 

experimentation began with a new technique that used alpha spectroscopy, which allowed the 

detection of the individual radionuclides in the sample. Alpha spectroscopy was used until 1988 

(Taylor 2000). 

Uranium Urinalysis

Analysis of uranium by measurement of emitted alpha radiation is referred to as an analysis for 

"enriched uranium." The analysis of uranium based on its chemical properties is referred to as a 

"depleted uranium" analysis. Enriched uranium (EU) was determined starting in the mid-1950s. 

Analyses for DU were performed with the ORNL fluorophotometric method from1954 to 1982. This 

method involved using a spectroscope to measure total uranium. The delayed neutron analysis 

(DNA) method was adopted for both EU and DU analyses around 1982 (Taylor 2000). 

Trivalent Actinide Urinalysis

The trivalent actinides consist of americium, curium, and californium. There was increased production 

of these materials at SRS in the early 1970s. Analysis was accomplished by extracting the trivalent 

actinides from the plutonium-tri-isooctlyamine (TIOA) raffinate with di,2ethylhexylophosphoric acid 

(HDEHP), gross alpha counting. According to a radiochemistry expert who reviewed the method, this 

procedure would also have captured elements higher than californium if present, specifically 

berkelium, fermium, and einsteinium. In 1990, ion exchange resin was used to separate trivalent

actinides from other actinides. The sample was direct-mounted (placed on a planchet with rimmed 

edges and heated to evaporation), and was "gross-alpha" counted by using a wide window on the 

alpha spectrometer. Since 1994, extraction chromatography resin has been used to separate trivalent 

actinides from other actinides. With the implementation of improved software and use of an 243Am

tracer, chemical recovery and counting efficiency can be determined for each sample (Taylor 2000).
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 18 of 36

Neptunium Urinalysis

Starting in 1959, neptunium was coprecipitated, ion exchanged, electrodeposited, and counted on 

Kodak film emulsion. In the mid-1960s (Butler 1968), the TIOA/gross alpha method was adopted. 

This method was replaced in 1993 by anion (negatively changed ions) exchange followed by direct 

mounting and gross-alpha counting.

Since 1994, extraction chromatography resin has been used to separate neptunium from fission 

products and other actinides, and electrodeposition has been used to mount the sample and gross 

alpha counted (Taylor 2000).

Fission Product Urinalysis

Fission products measurements in urine date to at least January 1957, although the method used at 

that time has not been discovered. Since 1959, select fission products in urine were analyzed by beta 

counting on a Geiger-Müller counter or proportional counter. The beta count was primarily used to 

determine the presence of 89Sr, 90Sr, and 90Y, which emit no gamma radiation. Because high-energy

betas are associated with 89Sr, a reliable measurement of its activity was made even though an

average of one gram of precipitate was mounted on the planchet. Yttrium-90, which emits a strong 

beta, also yields a reliable count, and affords a satisfactory estimate of its parent isobar, 90Sr, since 

some 90Sr betas are not counted due to absorption by the solid. The low energy beta of most of the 

induced activities along with the beta of zirconium-niobium and the rare earths will also contribute to 

the total gross beta count. No interference from 40K was observed. The chemistry on the beta 

sample resulted in quantitative recovery of beta-emitting isotopes of iron, cobalt, strontium, niobium, 

zirconium, ruthenium, and cerium. Gross gamma counting was performed using sodium iodide 

detectors. The gamma count was primarily used to determine the presence of induced activity. A 

gamma count was used because of the low counting efficiency of the weak beta energies associated 

with the induced activities. The gamma from zirconium-niobium and rare earths also contributed to 

the total gross gamma count. Either the beta or the gamma counting statistics are good enough to 

accurately estimate 1% of the MPA (maximum permissible amount in the total body) of any of the 

following radionuclides (i.e., 90Sr/Y, 89Sr, 95Zr/Nb, 95Nb, 144Ce/Pr, 60Co, 59Fe, 51Cr, 65Zn) with the 

exception of zirconium-niobium. The mixed beta-gamma isotope method provided a simple 

evaluation of the sample regarding the need for an extensive specific analysis, thus eliminating 

unnecessary and time consuming work (Boni undated).

SRS logbooks show that radionuclide-specific concentrations and associated uncertainties were 

quantified for some radionuclides for some of the gamma-counted samples starting in December 1965 

(DuPont 1965–1968). The reported radionuclides were:

• Cr-51

• Mn-54

• Co-60

• Zn-65

• Zr/Nb-95

• Ru-106

• I-131

• Cs-137

• Ba/La-140

• Ce-144.

Strontium Urine Bioassay

In the late 1950s and early 1960s, 89/90Sr was determined by beta counting on automatic and GeigerMueller counters along with zirconium, rare earths, and niobium. From 1969 to 1997, strontium was 

analyzed by a liquid ion-exchange method that extracted 90Y, which was counted on a beta 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 19 of 36

proportional counter. In 1998, the strontium method was modified by using an ion exchange column 

to extract 90Sr, which is also counted on a proportional counter. Table 5-1 lists the isotopes that were 

listed as being present, their radiations, bioassay method for detection, half-life, and the number of 

times they were identified in the monthly reports during the 1954–1971 period. An X indicates they 

were listed numerous times.

Table 5-1. Selected isotopes bioassay methods. 

Radionuclide

Times 

identified Half-lifea

Principal 

radiation

Bioassay method used or 

available for use

Bioassay 

period of 

availability

Am-243 X 7.4E3 yr Alpha, gamma Trivalent actinide urinalysis From mid 1960s

Whole-body counting of Np-239 From 1960

Chest counting From 1966b

Bk-249 X 320 d Beta None discoveredc

Co-60 X 5.26 yr Beta, gamma Whole-body counting From 1960

Gamma urinalysis IA From 1960

Cs-137 X 30.2 yr Beta, gamma Whole-body counting From 1960

Es-253 X 20.5 d Alpha Trivalent actinide urinalysisd From mid 1960s

Es-254 X 276 d Alpha Trivalent actinide urinalysisd From mid 1960s

Whole-body counting of Bk-250 

progeny

From 1960

Fm-254 X 3.24 h Alpha None discoverede

Pm-147 X 2.63 yr Beta None discoveredf

Pu-242 X 2.9E5 yr Alpha Pu urinalysisg From 1954

Tm-170h X 129 d Beta, gamma Chest countingi From 1966

Tm-171h X 1.92 yr Beta None discoveredj

U-233 X 1.6E5 yr Alpha EU urinalysis From 1959

Bk-250 2 3.22 hr Beta, gamma Whole-body countingk From 1960

Ce-144

2 284 d

Beta, gamma Fission product urinalysis FP From 1950s

Whole-body counting From 1960

Eu-152

2 12.7 y

Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Eu-154

2 8.8 yr

Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Tl-204 2 3.8 yr Beta Fission product urinalysis FP From 1950s

Am242ml 1 152 yr Beta (Am-242) Trivalent actinide urinalysis From mid 1960s

Ba-133h 1 10.5 yr Gamma Gamma urinalysis From 1960

Cm-248 1 3.4E5 yr Alpha Trivalent actinide urinalysis From mid 1960s

Es-255 1 38 d Alpha, beta 

(including progeny)

Trivalent actinide urinalysisd From mid 1960s

Np-239 1 2.35 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Pa-231 1 3.2E4 yr Alpha, gamma 

(7%)

Gamma urinalysis IA From 1960

Whole-body counting From 1960

Sb-124 1 60.2 yr Beta, gamma Gamma urinalysis IA From 1960

Whole body counting From 1960

Sb-125 1 2.73 yr Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Sc-46 1 83.8 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Tc-99 1 2.1E5 yr Beta Nonem

Y-91 1 58.8 d Beta Fission product urinalysis FP From 1950s

Ir-192 Shipment only 74.3 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Cd-113mh 0n 13.6 yr Beta Fission product urinalysis FP From 1950s

Cs-134h 0 2.06 yr Beta, gamma Fission product urinalysis FP From 1950s

Whole-body counting From 1960

Eu-155h 0 5.0 yr Beta, gamma Chest counting From 1966

(also whole-body counting of 

associated Eu-154)

From 1960
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 20 of 36

Radionuclide

Times 

identified Half-lifea

Principal 

radiation

Bioassay method used or 

available for use

Bioassay 

period of 

availability

Ho-166mh 0 1.2E3yr Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

I-131 0 8.06 d Beta, gamma Whole-body counting From 1960

Thyroid counting

K-40h 0 1.3E9 yr Beta, gamma Whole body countingo From 1960

Np-238 1 2.12 d Beta, gamma Whole-body counting if soon 

after intake

From 1960

a. Half-lives are taken from Lederer, Hollander, and Perlman (1967) and the Radiological Health Handbook (PHS 1970). 

b. The 40-cm arc whole-body detector had a lung calibration for the 98-keV gamma from Nd-147, which also would have 

detected the 99- 111-keV X-rays from Np-239.

c. See Attachment C.

d. The trivalent actinide method was reviewed by an expert radiochemist who confirmed that the elements higher than Cf 

(specifically Bk, Fm, and Es) would also have been captured. Alpha-emitting isotopes would have been counted.

e. The half-life of Fm-254 is so short that bioassay would not have been feasible. However, the cooling of the target 

before handling in the laboratory resulted in little or no leftover 254Fm. See Attachment C.

f. See Attachment B.

g. Pu-242 was used as an internal tracer starting in about 1981, but would have been detected in Pu urinalysis prior to 

that.

h. Isotopes that were discussed by Rupp (1969) in response to the request to use the SRS irradiation facilities for 

commercial and industrial development.

i. Because Tm is a bone seeker, the 84-kev gamma from the ribs would have been detectable in chest counts, despite the 

low yield (3.3%).

j. See Attachment D.

k. The half-life for Bk-250 is so short that the whole body count would have had to occur shortly after intake; however, the 

DAC (1E-7 μCi/cm3

) is 50 times larger than for Sr-90 so a significant intake would have been readily detectable in the 

workplace.

l. See Attachment E.

m. See Attachment A.

n. Not identified in the referenced monthly reports but were available along with bioassay methods for their detection.

o. An occupational intake would have been detectable only if the count was significantly larger than natural background.

Fecal Bioassay

Chemical bioassay of urine and fecal samples has been carried out at SRS since the 1950s to 

determine the individual's uptake of uranium and actinide elements such as americium, plutonium, 

and californium. Over the years, several techniques have been employed for these types of analyses. 

Methods ranging from autoradiography to high-resolution solid-state alpha spectroscopy have all been 

used at one time or another.

Many different methods of sample preconcentration, treatment, and preparation have been evaluated 

and used over the years. Each technique had some particular advantage at the time it was being 

used, either in terms of time required for analysis, the number of samples that could be processed 

simultaneously, the physical form of the final material to be counted, compatibility with other analyses 

that could be carried out sequentially or in parallel with the main analysis, compliance with more 

stringent detection levels, etc. (WSRC 1990). In contrast to a urine bioassay, a fecal bioassay can 

(and usually does) contain material that was never in the systemic body (i.e., insoluble material 

cleared from the lungs and gastrointestinal tract). Fecal bioassay is thus complementary to urine 

bioassay. In fact, at SRS, fecal bioassay is always used in conjunction with a urine bioassay and 

never alone (WSRC 2001). 

The primary method for fecal bioassay used in the early and mid-1980s was gamma spectrometry 

with a phoswhich detector. Some of the samples analyzed by this method were subsequently dried 

and sent to offsite laboratories for analysis. Since 1994, fecal samples have been analyzed by a 

modified urinalysis procedure (Taylor et al. 1995; Taylor 2000).
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In Vivo Bioassay - History of Whole-Body Counting at SRS (Taylor 2000) 

In vitro bioassay was the only method available for assessing occupational intakes of radionuclides in 

defense facilities like SRS before the mid-1950s. Before the development of large NaI detectors, in

vivo bioassay was performed with ionization chambers or Geiger-Mueller tubes that were rather 

insensitive and not practical for assessing occupational intakes of radionuclides, with the possible 

exception of 131I. In 1957, Sanders proposed that a whole-body counting facility similar to that at ANL 

using NaI detectors be built at SRS. Before the completion of the facility in 1960, only in vitro 

bioassay, primarily urinalysis, was performed at SRS. In vitro urine bioassay for fission and activation 

products performed during this period was typically a gross beta or gross gamma analysis.

The original SRS Whole Body Counter in Building 735-A was completed in 1960 and ceased 

operations in September 1995. Radionuclides counted were fission products and neutron activation 

products with a minimum energy of 135 keV from 144Ce and a maximum energy of 1,332 keV of 60Co. 

Sometime during the early 1970s, the original chair geometry was replaced with a recumbent (bed) 

geometry. The single 8-in, by 4-in. NaI detector was replaced with five 4-in, by 4-in. cylindrical NaI 

detectors under the bed. The NaI detectors were in an arc under the bed to provide constant 

efficiency for all detectors (Taylor 2000).

History of Chest Counting at SRS (Taylor 2000) 

The whole-body counter, which was useful for monitoring workers for intakes of gamma-emitting 

fission and activation products, was of no use for detecting or quantifying intakes of many actinides 

such as plutonium and americium. The 40-cm arc whole body counter was calibrated for 147Nd in the 

lung as well as in the whole body so low-energy gamma emitters (around 100 keV) could be 

quantified, although the sensitivity was poor. Efforts began at SRS in the mid-1960s to develop lowenergy photon detectors that would permit detection of actinides in the chest. Most efforts involved 

either thin NaI detectors or xenon-filled proportional counters. The xenon proportional counters, with 

their high resolution, showed great promise but were eventually abandoned because of their high

background and low efficiency. Development of the thin NaI detector continued, and in 1966 a 1-mmthick by 5-in.-diameter detector with a 0.001-in. window was first used to count workers involved in an 

incident. Thin NaI detectors were used with varying degrees of success from 1966 until 1971, when 

they were replaced with phoswich detectors. The results of the thin NaI counts were verified to be 

officially recorded in the personnel files of the workers counted during this time period. 

In the late 1960s, the phoswich detectors were developed for low-energy photon and X-ray 

measurement. Phoswich detectors consist of a "sandwich" of phosphorescent materials (NaI and 

CsI), hence, the name "phoswich". This detector took advantage of its thin construction and inherent 

anti-coincidence counting capabilities to greatly reduce the background in the low-energy region of the 

spectrum. In 1972, a dual phoswich system composed of two 5-in.-diameter detectors was placed in 

service. Although improvements were made in the analysis of data and calibration over the years, 

this basic system was used for chest counting for nearly 15 years until the adoption of planar 

germanium detectors in the late 1980s.

In the late 1970s, small coaxial germanium detectors were installed in the steel room. These 

detectors were pointed at the sides of the chest in an effort to detect the 186-keV photon from the 235U 

in enriched uranium, which was not readily detected by the phoswich or NaI detector. These 

detectors were also used to identify high-energy photon emitters, taking advantage of the superior 

resolution of the germanium detectors. In the mid-1980s, thick phoswich detectors specifically 

designed for enriched uranium were purchased. This turnkey system was used for only a few years 

before the germanium chest counter replaced it. In the early 1980s, solid-state planar germanium 

detectors were introduced at several facilities in the complex for chest counting. Routine use of a sixdetector germanium chest counter began at SRS in the 1987–1988 timeframe. This was a turnkey 

system purchased from a commercial vendor.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 22 of 36

The utility of routine chest counting for actinides declined throughout the 1980s and early 1990s. This 

decline can be related to fundamental changes in the occupational internal dose limits. During the 

1970s, the occupational limit for plutonium was expressed in terms of a quantity of plutonium that 

could be present in the chest at any given time – the maximum permissible lung burden (MPLB). The 

MPLB for plutonium was 16 nCi, which at the time was considered to be detectable with the dual

phoswich system. Thus, compliance with the occupational dose limits could be demonstrated with a 

simple chest count. In 1980, the limits changed to annual effective dose equivalent, which caused an 

increase in the complexity of calculations but did not cause insurmountable problems because it is 

closely related to the MPLB. The routine chest count was still a valuable tool for demonstrating 

compliance. In 1992, the limits were changed to committed effective dose equivalent. The effect of

this was to reduce drastically the amount of plutonium in the chest that corresponded to the 

occupational limit. The reduction was so great that chest counters no longer had any chance of 

detecting plutonium at compliance levels, rendering routine chest counts useless for determining 

compliance. However, chest counters still play an important role for special chest counts and for 

routine chest counts if a tracer such as 241Am might be present.

Germanium detectors have become the detectors of choice for measurement of actinide radioisotopes 

in the lungs. They are less efficient but much more specific (i.e., better resolution) than NaI or 

phoswich detectors. The SRS germanium detector system became operational in December 1989, 

replacing the phoswich detector system.

6.0 SUMMARY

Based on the various isotopes that were irradiated in the SRS reactors, it is clear that more than 100 

have been identified as having been irradiated for the SRL and many off-site customers. As indicated 

in the tables, aside from the “production” radioisotopes, most materials that were irradiated were 

relatively small quantities, which were, according to Gray (2006), sealed in aluminum cans on arrival, 

handled under water, and shipped as soon as possible on removal from the reactor. During the early 

years, urine bioassay (either gross alpha or gross beta analyses) was the only method of evaluating 

intakes of radioactive materials. SRS built a whole-body counter in 1960 that was used for fission and 

activation products. Because the whole-body counter had limited use for the detection of actinides, 

efforts began in the mid-1960s to develop low-energy photon detectors that would permit actinides to 

be detected in the chest. Improvements in whole-body counters and chest counters provided the 

means of detecting lower amounts of activities in the body. The Radiation Control Group provided 

training and retraining of all personnel and maintained records of internal and external exposures from 

the beginning of operations including exposures related to incidents. 

SRL did handle high activities of actinides, such as isotopes of americium, curium, and californium, for 

which bioassay was established. A program involving 210Po was conducted by a small group of 

personnel for whom bioassay was conducted (Mahathy 2011). Whole-body counting was used to 

monitor intakes of most of the exotic irradiated targets. The SRL did early development work 

involving chemistry on 171Tm and broken wafers (170Tm), which was one-time, small-scale work. 

Other than that, no evidence was found that the SRL worked with exotic radionuclide sources in a 

manner that breached the integrity of the source. Although this was small-scale work, a method for 

bioassay for the 170Tm was available as indicated in Table 5-1. No evidence was found for a bioassay 

for 171Tm, 249Bk, 254Fm, 147Pm, and 99Tc. These five radionuclides are discussed in Attachments A 

through E. There were exotic radionuclides created at SRS for which no bioassay was performed 

(mostly beta-only emitters), there was no evidence found that indicated that bioassay was needed. 

This report addresses 249Bk during the High-Flux demonstration period (1965-1970) (see Attachment 

C). However, a recently acquired document (Gibbs, 1994, p.19) indicates there was significant 

inventory of 249Bk from 1975 -1980. Additional research is required to identify its source during this 

period.
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 23 of 36

REFERENCES

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Plant, Aiken, South Carolina, July 9. [SRDB Ref ID: 49925]

Allen, H. F., 1964, Irradiation of Thulium, Erbium, Scandium, and Thallium, DPSOX05941, E. I. du 

Pont de Nemours and Company, Savannah River Plant, Aiken, South Carolina, July 19. 

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Boni, A. L., undated, Rapid Determination of Mixed Beta-Gamma Mixtures in Urine, DPSPU-58-030-

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Technical Department, DPSP-65-01-08, Savannah River Plant, Aiken, South Carolina, 

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DuPont (E. I. du Pont de Nemours and Company), 1965b, Progress Report, May 1965, Works 

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Samples,” logbook, Savannah River Site, Aiken, South Carolina, December 28, 1965, to July 

11, 1968. [SRDB Ref ID: 98497]

DuPont (E. I. du Pont de Nemours and Company), 1966a, Progress Report, January 1966, Works 

Technical Department, DPSP-65-1-1 DV, Savannah River Plant, Aiken, South Carolina, 

February 18. [SRDB Ref ID: 11879]

DuPont (E. I. du Pont de Nemours and Company), 1966b, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report July – September 1966, DP-1088-II, Savannah River 

Laboratory, Aiken, South Carolina, October. [SRDB Ref ID: 49310]

DuPont (E. I. du Pont de Nemours and Company), 1966c, Progress Report, March 1966, Works 

Technical Department, DPSP-66-1-3 Extract, Savannah River Plant, Aiken, South Carolina, 

April 20. [SRDB Ref ID: 68109]

DuPont (E. I. du Pont de Nemours and Company), 1966d, Radiation Control, Radiological and 

Environmental Sciences Division, Savannah River Laboratory Monthly Report May 1966, DP66-93-5, Savannah River Laboratory, Aiken, South Carolina, June 13. [SRDB Ref ID: 75711]

DuPont (E.I. du Pont de Nemours and Company), 1967a, Progress Report, March 1967, Works 

Technical Department, DPSP 67-1-3, Savannah River Plant, Aiken, South Carolina, April 20. 

[SRDB Ref ID: 12455]

DuPont (E. I. du Pont de Nemours and Company), 1967b, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report January – March 1967, DP-1105-II, Savannah River 

Laboratory, Aiken, South Carolina, October. [SRDB Ref ID: 48619]

DuPont (E. I. du Pont de Nemours and Company), 1967c, Radiation Control, Radiological and 

Environmental Sciences Division, Savannah River Laboratory Monthly Report December 

1966, DP-66-93-12, Savannah River Laboratory, Aiken, South Carolina, January 4. [SRDB 

Ref ID: 75740]

DuPont (E. I. du Pont de Nemours and Company), 1967d, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report April – June 1967, DP-1120-II, Savannah River 

Laboratory, Aiken, South Carolina, August. [SRDB Ref ID: 49317]

DuPont (E. I. du Pont de Nemours and Company), 1967e, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report June – September 1967, DP-1129-II, Savannah River 

Laboratory, Aiken, South Carolina, November. [SRDB Ref ID: 49343]

DuPont (E. I. du Pont de Nemours and Company), 1968a, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report October – December 1967, DP-1143-II, Savannah River 

Laboratory, Aiken, South Carolina, February. [SRDB Ref ID: 49344]

DuPont (E. I. du Pont de Nemours and Company), 1968b, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report January – March 1968, DP-1155-II, Savannah River 

Laboratory, Aiken, South Carolina, April. [SRDB Ref ID: 49345]
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DuPont (E. I. du Pont de Nemours and Company), 1968c, Savannah River Laboratory Isotopic Power 

and Heat Sources, Quarterly Report April – June 1968, DP-1169-II, Savannah River 

Laboratory, Aiken, South Carolina, August. [SRDB Ref ID: 49923]

DuPont (E. I. du Pont de Nemours and Company), 1970a, Curium Assimilation - SRL 

Decontamination Facility 773-A, SHI #273 Revised, Savannah River Laboratory, Aiken, South 

Carolina, April 20. [SRDB Ref ID: 45094, p. 322]

DuPont (E. I. du Pont de Nemours and Company), 1970b, Radiation Control Radiological Sciences 

Division Savannah River Laboratory Monthly Report December 1969, DP-69-93-12, Savannah 

River Laboratory, Aiken, South Carolina, January 7. [SRDB Ref ID: 48648]

DuPont (E. I. du Pont de Nemours and Company), 1970c, Radiation Control Radiological Sciences 

Division Savannah River Laboratory Monthly Report April 1970, DP-70-93-4, Savannah River 

Laboratory, Aiken, South Carolina, May 1. [SRDB Ref ID: 75802]

DuPont (E. I. du Pont de Nemours and Company), 1971a, Curium Assimilation, SHI #283, Savannah 

River Laboratory, Aiken, South Carolina, April 12. [SRDB Ref ID: 45094, p. 366]

DuPont (E. I. du Pont de Nemours and Company), 1971b, Works Technical Department Report for 

December, 1971, DPSP 71-1-12, Savannah River Laboratory, Aiken, South Carolina. [SRDB 

Ref ID: 68264]

DuPont (E. I. du Pont de Nemours and Company), 1971c, History of Large-Scale Transuranium 

Isotope Production at Savannah River, DPST-70-415 Part B (draft), Savannah River 

Laboratory, Aiken, South Carolina, February 26. [SRDB Ref ID: 72782]

DuPont (E. I. du Pont de Nemours and Company), 1971d, Works Technical Division Report for 

February 1971, DP-71-1-2 (extract), Aiken, South Carolina. [SRDB Ref ID: 68020]

DuPont (E. I. du Pont de Nemours and Company), 1974, History of the Savannah River Laboratory, 

Volume VI – Isotope Technology, DPST-62-200-VI, Savannah River Laboratory, Aiken, South 

Carolina, February. [SRDB Ref ID: 49498]

DuPont (E. I. du Pont de Nemours and Company), 1984a, History of the Savannah River Laboratory, 

Volume IV – General, DPST-62-200-IV, Savannah River Laboratory, Aiken, South Carolina, 

August. [SRDB Ref ID: 49497]

Du Pont (E. I. du Pont de Nemours and Company), 1984b, History of the Savannah River Laboratory, 

Volume II – Separations Technology, DPST-62-200-II DEL, Savannah River Laboratory, 

Aiken, South Carolina, June. [SRDB Ref ID: 89523]

DuPont (E. I. du Pont de Nemours and Company), 1984c, History of the Savannah River Laboratory, 

Volume I – Production Reactor, Fuel, and Target Technology, DPST-62-200-I DEL, Savannah 

River Laboratory, Aiken, South Carolina, August. [SRDB Ref ID: 89232]

Gibbs, A, 1994, Uranium, Thorium, and TRU Inventories at SRS (U), WSRC-TR-940255 

Westinghouse Savannah River Company, Savannah River Site, Aiken, South Carolina. 

[SRDB Ref ID: 47576]

Gray, P., 2006, “High Flux @ SRP,” e-mail to D. Bihl (Oak Ridge Associated Universities Team), April 

29. [SRDB Ref ID: 23544] 
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Harbour, R. M., C. H. Ice, W. H. Hale, and J. T. Lowe, 2000, “Development of Chemical Processes 

and Equipment to Recover Curium-244 and Californium-252,” Proceedings of the Symposium, 

May 17, 2000, Fifty Years of Excellence in Science and Engineering at Savannah River Site, 

WSRC-MS-2000-00061, Westinghouse Savannah River Company, Savannah River Site, 

Aiken, South Carolina. [SRDB Ref ID: 45046, p. 168] 

La Bone, T. R., 2011, Kinetics of 242mAm and its Daughters, Oak Ridge Associated Universities Team, 

Oak Ridge, Tennessee, November 8. [SRDB Ref ID: 104330] 

Isotope Charts, 2011, Isotope Charts for the Years 1953 Through 1976, Savannah River Site, (SRDB 

ID: 94568)

Lederer, C. M., J. M. Hollander, and I. Perlman, 1967, Table of Isotopes, 6th Edition, John Wiley and 

Sons, New York, New York.

Mahathy, M. M., 2011, Feasibility of Reconstructing Doses Received from Potential Exposures to 

Polonium at Savannah River Site, ORAUT-RPRT-0048 Rev 00, Oak Ridge Associated 

Universities Team, Oak Ridge, Tennessee, January 24. 

PHS (Public Health Service), 1970, Radiological Health Handbook, PHS-PUBL-2016, U.S. 

Department of Health, Education, and Welfare, Consumer Protection and Environmental 

Health Service, Rockville, Maryland, January. [SRDB Ref ID: 75017]

Reed, M. B., M. T. Swanson, S. Gaither, J. W Joseph, and W. R. Henry, 2002, Savannah River Site at 

50, U.S. Government Printing Office, Washington, D.C. [SRDB Ref ID: 24750] 

Rupp, A. F., 1969, “Radioisotope Production in SRO High Flux Runs,” letter to E. E. Fowler (U.S. 

Atomic Energy Commission, Division of Isotopes Development), Union Carbide Corporation, 

Nuclear Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, August 22. [SRDB 

Ref ID: 13162, pp. 2–6] 

Sanders, Jr., S. Marshall, W. J. Kerrigan, E. L. Albenesius, 1969, Radiation Shielding of Small Power 

Sources of 170,171Tm, DP-1158, Savannah River Laboratory, Aiken, South Carolina, January. 

[SRDB Ref ID: 48622]

SC&A (Sanford Cohen & Associates), 2010, List of Exotic Radionuclides, December 10. [SRDB Ref 

ID: 94568] 

Smith, J. A., E. J. Hennelly, C. H. Ice, and H. P. Allen, 1965, “Isotope Production in a Savannah River 

Reactor at Flux Levels above 2015 n/(cm2

)(sec),” The Savannah River Flux Demonstration, 

DP-999, compiled by J. L. Crandall, E. I. du Pont de Nemours and Company, Savannah River 

Laboratory, Aiken, South Carolina, June. [SRDB Ref ID: 47396, pp. 21–30] 

Taylor, G. A., 2000, The Evolution of Internal Dosimetry Bioassay Methods at the Savannah River 

Site, WSRC-MS-2000-00290, Westinghouse Savannah River Company, Savannah River Site, 

Aiken, South Carolina. [SRDB Ref ID: 46146] 

Taylor, G. A., K. W. Crase, T. R. LaBone, and W. H. Wilkie, 1995, A History of Personnel Radiation 

Dosimetry at the Savannah River Site, WSRC-RP-95-234, Westinghouse Savannah River 

Company, Savannah River Site, Aiken, South Carolina, May. [SRDB Ref ID: 12098]
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Till, J. E., J. W. Aanenson, P. J. Boelter, M. J. Case, M. Dreicer, H. A. Grogan, M. O. Langan, P. D. 

McGavran, K. R. Meyer, R. Meyer, H. J. Mohler, A. S. Rood, R. C. Rope, S. K. Rope, L. A. 

Stetar, P. G. Voilleque, T. F. Winsor, and W. Yang, 2001, Savannah River Site Dose 

Reconstruction Project, RAC Report No, 1-CDC-SRS-1999-Final, Risk Assessment 

Corporation, Neeses, South Carolina, April 30. [SRDB Ref ID: 11259]

WSRC (Westinghouse Savannah River Company), 1990, Internal Dosimetry Technical Basis Manual, 

WSRC-IM-90-139, Savannah River Site, Aiken, South Carolina. [SRDB Ref ID: 11266] 

WSRC (Westinghouse Savannah River Company), 2001, Internal Dosimetry Technical Basis Manual, 

WSRC-IM-90-139, Rev. 8, Savannah River Site, Aiken, South Carolina. [SRDB Ref ID: 722]
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 28 of 36

In the latter half of 1964 and into 1965 work was performed in the Savannah River Laboratory on 

extraction of 99Tc from high level waste (DuPont 1984b, pp. 94, 118; DuPont 1964b, p. 44). The feed 

material would have had plenty of fission products and monitoring by fission product urinalyses and 

whole body counting would have been appropriate. 1.1 g of 99Tc was isolated in June 1964 and work 

toward extracting another 1.8 g was conducted in September (DuPont 1964a, p. 44). The purity of the 

product was not mentioned. No mention of Tc extraction was made for subsequent years in the 

historical summary document (DuPont 1984b) and no mention was made in the series of quarterly 

reports of the Isotopic Power and Heat Sources Project starting in July 1966 (for instance, DuPont 

1966b, 1967b, and others). [Note: the October – December 1966 report has not been captured.] 

Additionally, no mention of work with 99Tc was made in any of the Savannah River Laboratory

Radiation Control Group monthly reports in 1966 (whereas there is frequent mention of work 

specifically with curium, thorium, 238Pu, 237Np and occasional mention of work with thulium and 

polonium in 1966 or other years). 

The work with 99Tc extraction appears to be limited from about midyear 1964 to about midyear 1965, 

when it appears work on this radionuclide was abandoned.

Because the specific activity of 99Tc is small (0.017 Ci/g), the potential for consequential intakes were 

smaller than for the other heat source radionuclides and were very small in terms of potential for 

inducing cancer. Table A-1 shows potential chronic intakes and committed organ doses from 1.8 g of 

99Tc using the common Brodsky’s “magic” number (10-6) to relate activity in process to potential for 

intakes (Brodsky 1980). 

ATTACHMENT A

TECHNETIUM-99

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 Table A-1. Potential committed organ doses

 (mrem) from 1.8 g 99Tc.

a

Target organs

Organ committed doses 

(mrem)

Chronic intake (365 d), 

83.7 pCi/d

Type F 

inhalation

Type M 

inhalation

Adrenals 3.86E-03 3.47E-03

Urinary bladder 1.60E-02 1.44E-02

Brain 3.86E-03 3.47E-03

Breast 3.86E-03 3.47E-03

Gall bladder 3.86E-03 3.47E-03

Heart wall 3.86E-03 3.47E-03

Kidneys 3.86E-03 3.47E-03

Liver 5.14E-03 4.63E-03

Muscle 3.86E-03 3.47E-03

Ovaries 3.86E-03 3.47E-03

Pancreas 3.86E-03 3.47E-03

Testes 3.86E-03 3.47E-03

Thyroid 1.01E-01 9.07E-02

R.B.M. 3.86E-03 3.47E-03

Bone surface 3.86E-03 3.47E-03

Stomach 2.07E-01 1.89E-01

S.I. 5.18E-03 6.24E-03

U.L.I. 5.59E-02 5.99E-02

L.L.I. 1.56E-01 1.69E-01

Skin 3.86E-03 3.47E-03

Spleen 3.86E-03 3.47E-03

Thymus 3.86E-03 3.47E-03

Uterus 3.86E-03 3.47E-03

ET 1.92E-01 9.33E-01

Lung 6.10E-03 2.69E+00

Colon 9.94E-02 1.07E-01

Esophagus 3.86E-03 3.47E-03

Gonads 3.86E-03 3.47E-03

a. Maximum doses highlighted. 

ATTACHMENT A

TECHNETIUM-99

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Promethium-147 was among the unusual nuclides that were produced in research quantities and 

evaluated for a variety of uses from 1965-1970. Promethium-147 has a half-life of 2.6 years and is a 

beta emitter with an average energy of 62 keV. The low energy allows for ease of handling and does 

not require significant shielding, which, in turn, reduces the weight required of other radioisotopes for 

similar applications. 

As early as 1958 there was an interest in using the fission product 147Pm as a fuel for SNAP batteries. 

As a result, a process for the recovery of 147Pm was started. At that time a method was also 

developed for the radiochemical determination of 147Pm and for 144Ce and 91Y, which were the major 

contaminants of promethium (DuPont 1984b, p. 23). As can be seen in Table 3-4 147Pm was 

identified several times in 1959 and 1960. However, it was not until 1965 that studies began on the 

use of various radionuclides as heat source materials, including 147Pm. Special irradiations were 

made to produce test quantities of 170Tm, 171Tm, 46Sc, and 204Tl to measure the properties of these 

isotopes and to predict the feasibility of their large-scale production. An estimate was made of the 

annual production at SRP (in the fission-product fraction) of 147Pm, and the recovery of 147Pm from 

current high- activity waste was studied. The isotopic purity of aged 147Pm (from the SRP waste 

tanks) was also measured (DuPont 1974, pp.5-6).

Because NASA had an interest in aged 147Pm as a source of isotopic heat for space missions, a study 

of the quantity and quality of 147Pm that could be produced annually at SRP was undertaken. The 

major source of 147Pm for its recovery would be fission product waste. The quantity available after 

aging to allow for decay of associated 148Pm would be about 1700 W(t) per reactor-year of operation. 

Product quality from current waste in 1965, as represented by a low ratio of 148Pm/147Pm, would 

presumably be the same as that produced at Hanford (about 5E-7). Promethium 148 emits gamma 

rays that could require extra shielding while 147Pm is a beta emitter as indicated above. In addition, a 

scoping study was made to determine the quality of 147Pm in fission product sludge from SRP 

operation. The measured isotopic ratio of 148Pm in 147Pm was found to be similar to the isotope ratio 

in Hanford fission products (0.4 ppm) (DuPont 1974, p. 7).

A process to recover 147Pm from plant high activity waste was also developed in the laboratory in 

1965 for use as a power source. During the extraction process, the 147Pm has to be separated from 

144Ce since that is a major source of heat and its removal would allow easier storage of the 147Pm. 

After separation the 147Pm was then stored until most of the 148Pm impurity decayed, and was then 

purified by ion exchange (DuPont 1984b, p.119).

A plant test to demonstrate the initial recovery steps was begun in November 1966 and demonstrated 

the recovery of a 147Pm-lanthanide mixture from plant waste (DuPont 1984b, p. 142). In 1967 another 

test of the process to recover 147Pm from high activity waste was successful in producing about 500 g 

from a lanthanide mixture. Pressurized ion exchange columns were tested with non-radioactive 

solutions to isolate the 147Pm from the lanthanide mixture (DuPont 1984b, p. 163).

For another application suitable-isotopic purities were calculated for 238pu, 171Tm, and 147Pm as power 

sources for pumps to assist or replace the heart. A heart pump would be ~ 5 inches in diameter and 

installed in the abdomen. It would include a steam boiler and heat exchanger with blood as coolant

(DuPont 1968a, p. 45)). Plutonium-238 appeared to be the most promising power source for such 

pumps because it would be available in adequate supply and could be aged to remove 236Pu to 

sustain long cycles of operation with reasonable weight of shielding. Thulium-171 would require 

potentially the least shielding, but is expected to be more expensive than 238Pu in the necessary, 

highly purified form and is also handicapped by much shorter half-life (l.9 vs. 87 yr). Promethium-147, 

ATTACHMENT B

PROMETHIUM-147

Page 1 of 2 
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 31 of 36

readily available as a fission product, had similar disadvantages of being difficult and more expensive 

to produce the necessary isotopic purification required for minimum shielding and it also has a shorter 

half-life (2.6 yr). In 1968 147Pm was also considered as a heat source for thermoelectric and 

thermionic power generation. Studies of compounds of promethium were conducted to provide 

information on the optimum fuel form selection (DuPont 1984b, p.182). A plant test of the process 

recovered about 2 kg of 147Pm as a lanthanide mixture. In laboratory tests pressurized ion exchange 

columns were used to isolate approximately 1 g of 147Pm from a portion of the lanthanide mixture 

(DuPont 1984b, p. 182).

While there was considerable development work with 147Pm there is no record that indicates it was 

ultimately applied to the applications mentioned above. However, the separation of 500 g and 2 kg in 

different operations represents multicuries of 147Pm activity for which no bioassay method has been 

identified.

In 1969 work on the development of 147Pm as a heat source was concluded and no further work was 

planned (DuPont 1984b, pp. 193, 200 & 201 ) and in 1971 about 460 g 147Pm and 100 g 144Ce, stored 

for several years, were discarded to make room for 244Cm (DuPont 1971b, p. 8). 

ATTACHMENT B

PROMETHIUM-147

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Berkelium-249 and 254Fm were two of many radionuclides created as part of the Transplutonium 

Project, evolving out of the work to create 252Cf and to produce 3 kg of 244Cm by 1967 (DuPont 1964a, 

p. 25). Berkelium-249 resulted from decay of 249

Cm, which was produced from multiple (n,γ) reactions 

starting from 244Cm. The Savannah River Laboratory monthly report for September 1964, in a section 

on the use the tramex process for isolation of Cf, stated “Continuing analytical development study will 

include determination of Bk and Es, in addition to Cf, in the dissolver solution from the second 

Transplutonium campaign” (DuPont 1964b, p. 34). There was no mention of Bk in the October and 

November monthly reports for 1964. “A series of laboratory tests was made in support of the second 

Transplutonium separations campaign, which was started in the H Area in October.” These tests 

focused on the separations chemistry of Pu, Am, and Cm (DuPont 1964c, p.38). 

Fm-254 resulted from multiple (n, γ) reactions starting from 252Cf with the final two steps being an (n,γ) 

reaction of the 253Es and beta decay of 254Es (DuPont 1971c, p. 38). Fm-254 (half-life 3.2 hr) was an 

intermediate step in the pathway to produce 257

Fm (half-life 84 d). The same historical document 

states that “only 257Fm can be recovered in significant amounts because, even though it is at the end 

of the reaction chain, its 84-day half-life is much longer than those of the other fermium isotopes” 

(DuPont 1971c, p. 39). 

The Savannah River Laboratory monthly reports for the second half of the 1960s have not been 

captured. The historical summary document of the Separation Technology Section indicated that in 

1965 analyses were developed for 249Bk and 253Es as part of the Transplutonium Project (DuPont 

1984b, pp. 111, 112). However, actual separation of 249Bk is not mentioned until 1969 (DuPont

1984b, p. 199). The chapter in the fifty-year symposium document that discusses creation of curium 

and californium mentions a broader time period: “In 1968-69, a high-pressure elution development 

cation exchange chromatographic process was developed . . . Milligram quantities of 252Cf, 

submilligram quantities of 249Bk, and subnanogram amounts of 253Es and 254Fm were separated from . 

. . 244Cm and 243Am. . .” (Harbour et al. 2000, p. 172). In this time period SRS was also irradiating 

targets that were shipped to ORNL for separation of the Bk, Fm, and Es (DuPont 1971c, pp. 39, 40). 

In April 1970, tests at the Savannah River laboratory demonstrated the feasibility of isolating 249Bk, 

254Es, and 257Fm coproduced with 252Cf in SRP high-flux irradiations. Berkelium is eluted after 

californium and before curium . . . (DuPont 1971c, pp. 37 - 40). These tests were referred to as 

involving tracer levels. This process was apparently intended to be used in the Multipurpose 

Processing Facility, under construction in 1970 in the F Area hot canyon (DuPont 1984b, p. 209, 210). 

The Separation Technology Section history indicates that work on isolating these elements continued 

into 1971 (DuPont 1984b, p. 225). This may have been in conjunction with the Californium I 

campaign. The Transplutonium Project history document, drafted in February 1971, indicates that 

“Californium I campaign is expected to yield 51 milligrams of 249Bk, 6 micrograms of 254Es, and 10 

picograms of 257Fm ( DuPont, 1971c, p. 40). Note that no mention was made of 254Fm. There are 

conflicting reports on the time of the Californium I campaign. The February 1971 document refers to it 

in the future and other documents show Cf work underway in early 1971 (DuPont 1971d, pp. 5, 6, 8, 

13); however the reactor history states that the Californium – I campaign occurred from August 1969 

to November 1970 (DuPont 1984c, p. 158). The difference might be time the targets were in the 

reactor versus the time the targets were being analyzed after a cooling period. The later article on 

this work (circa 2000) refers to submilligram quantities of 249Bk not tens of milligrams (Harbour et al. 

2000, p. 172) so it appears that the predicted amounts in the Californium 1 campaign were either not 

created or not recovered in the amounts expected. 

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

Page 1 of 2
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 33 of 36

There is no indication that the elements higher than Cf were actually isolated in the Multipurpose 

Processing Facility when it was operating, circa 1972. Savannah River Laboratory monthly reports in 

1972 and 1973 indicate that production was focused on 238Pu heat sources and 252Cf neutron sources.

From an intake perspective, it is important to note concerning general contamination in glove boxes, 

caves, accessible surfaces, and air contamination, the overwhelming constituents would have been 

244Cm, 252Cf, and 243Am, which were being handled in large amounts. For instance just disposal of 

244Cm contaminated waste to the burial grounds in 1969 totaled 12,185 Ci (DuPont 1970b, p. 8). The 

Radiation Control Group monthly reports show that handling of 252Cf in the hundreds of Ci range in the 

Savannah River Laboratory was common during this period. Possible exposure to pure submilligram 

quantities of 249Bk (1.6 Ci/mg) or the microgram/picogram quantities of einsteinium and fermium would 

have only occurred to those handling and packaging the final products of the chemical process, which 

would have been handled with extreme care because of the rarity of the products. For instance, 

there is no mention of contamination by Bk, Es, or Fm in the April 1970 Radiation Control Group 

report (DuPont 1970c) (when the tests mentioned in the preceding paragraph were conducted), or in 

any of that series of reports in 1969 or 1971. 

Possible exposure to pure 249Bk appears to have occurred intermittently during 1969 through 1971 

and maybe in 1968. It is likely the dates and amounts could be determined more precisely by capture 

of the Savannah River Laboratory monthly reports during these years. However, general 

contamination in the workspace would have been dominated by other radionuclides and the risk of 

contamination to the small amounts of pure 249Bk seem low. 

Both 254Fm and 257Fm are alpha emitters so, if present in the body, would have been excreted in urine 

and in theory would have been measured by the trivalent actinide urinalysis. However, because of its 

3.2-hr half-life, essentially none of the excreted 254Fm would have been counted considering the time 

gap between excretion and counting of the sample. However, the same half-life consideration applies 

to time involved in separating and extracting the fermium from the californium and einsteinium. For 

instance, the 1971 history indicates that a 15-month cooling period between the time the targets are 

removed from the reactor until the start of chemical processing is normal (DuPont 1971c, p. 40). If 

hardly any cooling time was employed then no 254Fm would be left in the target (the parent 254Es also 

has a half-life in hours). Although the article in 2000 states “subnanogram amounts of . . . 254Fm were 

separated” (Harbour et al. 2000, p. 172), that article does not mention 257Fm, which was the actual 

goal of the irradiation of einsteinium. It is possible that this was a typo and the sentence should have

said 257Fm. The purpose of creating the higher order elements was for research on their physics and 

chemical natures, which would not have been possible on a very small amount of something with a 

3.2-hr half-life. It is concluded that no exposure occurred to 254Fm as a pure radioisotope.

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

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Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 34 of 36

Interest in 171Tm (and 170Tm) as potential heat sources seriously started in 1964. In the September 

monthly report of the Savannah River Laboratory, under New Products – Special Test Irradiations, 

“quantities of 46Sc, 170Tm, 171Tm, and 204Tl are being made in the current 244Cm-producing charge” 

(DuPont 1964a, p.43), a test that was authorized in July (Allen 1964). The October and November 

monthly reports do not mention any actual handling of the material (DuPont 1964b,c). An historical 

summary document also mentions the above irradiation in 1964 and states that an ion exchange 

process for separating the thulium from the erbium target has been developed and will be used on the 

irradiated material (DuPont 1984a, p. 93). 

The first chemistry to extract 170Tm and 171Tm from the irradiated erbium targets occurred in 1965 

using the aforementioned technique. This was at the laboratory development phase (DuPont 1984a, 

p. 118). Experimentation and improvement of this technique continued in 1966 (DuPont 1984a, p. 

141). During this period the material extracted consisted of a combination of 170Tm and 171Tm 

because the target material was natural erbium. The Radiation Control Group monthly report for May 

indicates that a study on the radiation characteristics from and shielding design needed for 170,171Tm 

sources was underway (DuPont 1966d, p. 8). Because 170Tm is a strong beta-emitter, intakes of this 

mixed isotopic material would have been detected by both air sampling and the fission product 

bioassay.

The Radiation Control Group December 1966 monthly report indicated that proposed operations for 

about 100 Ci of 171Tm in glove boxes B-126 – 130 were approved, and that the associated beta and 

low-energy gamma activity was to be easily shielded. The report added that the aluminum cladding 

was to be removed in the High Level Caves (DuPont 1967c, p.10).

By second quarter of CY 1967 the production method had evolved to irradiation of erbium enriched to 

96% 170Er (DuPont 1967d, p. 21, DuPont 1984a, p. 163). This produced 171Tm directly with little 

contamination with 170Tm. Two hundred Ci of 171Tm had been produced by this method at that time. 

In the following quarter about 60 Ci of this material was purified from the residual erbium using cation 

exchange chromatography (DuPont 1967e, p. 22). No work with 171Tm was mentioned in the fourth 

quarter CY 1967 report. That report indicated that 238Pu would be a better heat source than either 

171Tm or 147Pm because of their short half-lives, difficulty in separation and purification, and cost 

(DuPont 1968a, p. 45). The first quarter CY 1968 report only discusses work with 244Cm and 170Tm; 

external radiation measurements were made on a 3 kCi 170Tm source that also contained 1.7 kCi of 

171Tm but no chemistry was done (DuPont 1968b, p. 29; (Sanders et al. 1969). The second quarter 

CY 1968 report states that another separation of 171Tm from enriched Er was made with improved 

chemistry methods (30 mg 171Tm, 57 Ci) (DuPont 1968c, pp. 45-46). That report also stated that this 

concluded the work with 171Tm. That the work was “deferred” was reiterated in a July 1969 letter from 

E. L. Albenesius, Research Manager in the Savannah River Laboratory (Albenesius 1969). No further 

work with potential for intakes of 171Tm was found.

No mention of the work with 171Tm was found in the series of monthly reports by the Radiation Control 

Group of the Savannah River Laboratory in 1968 or 1969, which merely indicates that there was no 

significant loss of control. 

Bebbington, in his history of DuPont years at Savannah River, concluded “among the unusual 

radionuclides produced in research quantities and evaluated as heat sources and for a number of 

other specialized uses were: curium-244, promethium-147, thulium 170 and 171, scandium-46, 

thallium-204, and polonium-210. . . . . None of these nuclides, though, found significant applications” 

(Bebbington 1990, p. 200).

ATTACHMENT D

THULIUM-171

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Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 35 of 36

The period of work with potential for intakes of essentially pure 171Tm appears to be limited to 1967 

and 1968. The work took place in the Savannah River Laboratory. 

ATTACHMENT D

THULIUM-171

Page 2 of 2
Document No. ORAUT-RPRT-0054 Revision No. 00 Effective Date: 01/20/2012 Page 36 of 36

Am-242m is a beta emitter with a half-life of 152 years. It has an alpha decay branch but the yield is 

only 0.5%. Because it does not emit significant alpha radiation it would not be counted in the trivalent 

actinide bioassay method. The principal decay branch of 242mAm (99.5% yield) beta-decays to 242Am 

which beta-decays to 242Cm with an 82.7% yield and 16-hour half-life. So this latter transition is nearly 

instantaneous, and equilibrium between 242mAm and 242Am is rapidly established. Curium-242 is an 

alpha-emitter with a 163-day half-life that would have been counted by the trivalent actinide bioassay 

analysis. However, it takes quite a few years for equilibrium to be established between 242mAm and 

242Cm and equilibrium is never 100% because of the branching ratios. 

The question is whether the buildup of 242Cm is rapid enough and sufficient such that the trivalent 

actinide bioassay would have been an appropriate bioassay for 242mAm exposure. Because 

americium and curium behave the same in the body, body biokinetics do not change the ratio of 242Cm

and 242mAm as a function of time. Assuming workers were chronically exposed to initially pure 242mAm, 

the rate at which 242Cm activity would build up in the body is shown in Table E-1, assuming a unit 

activity intake per day of 242mAm.

Table E-1. Activity of 242Cm per 

unit daily intake of 242mAm. 

Time after start of 

chronic intake

Cm-242 

activity

1 d 0.0018

5 d 0.0087

10 d 0.017

50 d 0.082

90 d 0.140

180 d 0.249

1 yr 0.407

1.5 yr 0.504

2 yr 0.571

3 yr 0.648

The derivations and a plot of the buildup as a function of time are shown in La Bone (2011). It is 

unlikely that the 242mAm would have been pure in the workplace environment; i.e., airborne or surface 

contamination would have been a mix of 242mAm and 242Cm so the actual ratios would have been 

larger than those in Table E-1, up to an equilibrium of 0.83. The activity values in Table E-1 show that 

an annual or biennial trivalent actinide bioassay would have had sufficient 242Cm to monitor for chronic 

intakes of 242mAm.

ATTACHMENT E

Page 11 of 1 

ATTACHMENT E

AMERICIUM-242M